CONCEPTUAL DESIGN FOR A RE-ENTRANT TYPE FUEL CHANNEL FOR SUPERCRITICAL WATER-COOLED NUCLEAR REACTORS

Transcription

1 CONCEPTUAL DESIGN FOR A RE-ENTRANT TYPE FUEL CHANNEL FOR SUPERCRITICAL WATER-COOLED NUCLEAR REACTORS by Jeffrey Samuel A Thesis Submitted in Partial Fulfillment of the Requirements for the Degree of Master of Applied Science in The Faculty of Energy Systems and Nuclear Science University of Ontario Institute of Technology April, 2011 Jeffrey Samuel, 2011

2 ABSTRACT Current CANDU-type nuclear reactors use a once-through fuel-channel with an annulus gas insulating it from the moderator. The current reference design for a CANDU-type SuperCritical Water-Cooled Reactor (SCWR) is to eliminate the annulus gap and use a ceramic insert to insulate the coolant from the moderator. While such a design may work, alternative fuel-channel design concepts are under development to explore the optimum efficiency of SCWRs. One such alternative approach is called the Re-Entrant fuel-channel. The Re-Entrant fuel-channel consists of three tubes, the inner tube (flow tube), pressure tube and an outer tube. The fuel bundles are placed in the inner tube. An annulus is formed between the flow and pressure tubes, through which the primary coolant flows. A ceramic insulator is placed between the pressure tube and the outer tube. The coolant flows through the annulus receiving heat from the inner tube from one end of the channel to another. At the far end, the flow will reverse direction and enter the inner tube, and hence the fuel-string. At the inlet, the temperature is 350 C for a high-pressure coolant (pressure of 25 MPa), which is just below the pseudocritical point. At the outlet, the temperature is about 625ºC at the same pressure (the pressure drop is small and can be neglected). The objective of this work was to design the Re-Entrant channel and to estimate the heat loss to the moderator for the proposed new fuel-channel design. A numerical model was developed and MATLAB was used to calculate the heat loss from the insulated Re-Entrant fuel-channel along with the temperature profiles and the heat transfer coefficients for a given set of flow, pressure, temperature and power boundary conditions. Thermophysical properties were obtained from NIST REFPROP software. With the results from the numerical model, the design of the Re-Entrant fuelchannel was optimized to improve its efficiency. ii

3 ACKNOWLEDGEMENTS Financial support from the NSERC/NRCan/AECL Generation IV Energy Technologies Program and NSERC Discovery Grants is gratefully acknowledged. I am deeply grateful to my supervisors, Dr. Harvel and Dr. Pioro, for their support and encouragement in my research and the writing of this thesis. I sincerely appreciate their constant guidance and support during the period of this work and most of all, their patience and understanding. I would like to thank my colleagues Amjad Farah, Adepoju Adenariwo, Harwinder Thind, Lisa Grande and Graham Richards for their valuable discussions. I m also indebted to the members of the Nuclear Design Laboratory at UOIT for giving me the perfect working atmosphere and also for their kindness. I m especially thankful to my friends Ciandra D Souza, Krista Nicholson and Donald Draper, for helping me through the difficult times by keeping me sane, and for being a source of inspiration during the course of my research. Finally, I would like to thank my parents, Naomi and James Richard, for always being a tremendous source of emotional, moral and material support. iii

4 TABLE OF CONTENTS 1 INTRODUCTION Objective 5 2 BACKGROUND AND LITERATURE REVIEW Generation IV Nuclear Technology Generation IV Concepts SuperCritical Water-cooled Reactor Concepts SCW Properties SCW Correlations Fuel Channel Design Concepts 27 3 PROPOSED FUEL CHANNEL DESIGN CONCEPT 30 4 NUMERICAL MODEL OF RE-ENTRANT FUEL-CHANNEL Fundamental Equations Flow Area, Wetted Perimeter, Hydraulic Diameter Nodalization Initial Estimate of Temperature profiles for Hot and Cold Side Initial Estimate of Temperature of Fuel Sheath Thermal Resistances Heat Transferred from Hot Side to Cold Side and Heat Loss to Moderator from Cold Side Actual Temperature Profile of Cold Side, Hot Side & Fuel Sheath Surface Temperatures of Tubes 55 iv

5 4.10 MATLAB code 55 5 HEAT TRANSFER ANALYSIS Reference Case: No Heat Loss to the Moderator Heat Loss to the Moderator Non-insulated Re-Entrant channel Porous Yttria Stabilized Zirconia Insulated Re-Entrant Channel Zirconium Dioxide (ZrO 2 ) Insulated Re-Entrant Channel Heat Loss Comparison Impact of Non-Uniform Flux Shapes 93 6 CONCLUDING REMARKS FUTURE WORK 108 REFERENCES 109 APPENDIX A: NUMERICAL MODEL IN MATLAB 113 APPENDIX B: VERIFICATION OF NUMERICAL MODEL 127 APPENDIX C: IMPACT OF INSULATORS THERMOPHYSICAL PROPERTIES 128 APPENDIX D: IMPACT OF NON-UNIFORM FLUX SHAPES THERMOPHYSICAL PROPERTIES 132 APPENDIX E: CONTRIBUTIONS TO KNOWLEDGE 142 v

6 LIST OF TABLES Table 3.1: Parameters of fuel bundle options for the Re-Entrant channel 37 Table 3.2: Reference case Re-Entrant Channel Dimensions 40 Table 4.1: Reference case Re-Entrant Channel Parameters 46 Table 5.1: Heat loss comparison 78 Table 5.2: Heat loss comparison for insulated Re-Entrant channel 92 Table 5.3: Heat loss comparison for insulated Re-Entrant channel with different axial power profiles Table 5.4: Peak sheath temperature and location of peak sheath temperature for insulated Re-Entrant channel with different axial power profiles Table 5.5: Location of pseudocritical point in the Re-Entrant channel with different axial power profiles vi

7 LIST OF FIGURES Figure 2.1: Nuclear Reactor Technology in Canada 8 Figure 2.2: Pressure Vessel Type SCWR 13 Figure 2.3: Pressure Tube Type SCWR 13 Figure 2.4: T-s diagram for direct cycle SCWR with no-reheat option 14 Figure 2.5 (a): Specific Heat vs. Temperature of water in the pseudocritical region Figure 2.5 (b): Density vs. Temperature of water in the pseudocritical region 17 Figure 2.5 (c): Thermal Conductivity vs. Temperature of water in the pseudocritical region Figure 2.5 (d): Prandtl Number vs. Temperature of water in the pseudocritical region Figure 2.5 (e): Dynamic Viscosity vs. Temperature of water in the pseudocritical region Figure 2.5 (f): Kinematic Viscosity vs. Temperature of water in the pseudocritical region Figure 2.6: Comparison between SCW correlations when bulk fluid temperature is 350 C and wall temperature is 400 C Figure 2.7: CANDU type fuel-channel Figure 2.8: SCWR type fuel-channel concept 28 Figure 3.1 (a): Horizontal Re-Entrant channel configuration 32 Figure 3.1 (b): Horizontal Re-Entrant channel configuration 32 Figure 3.2 (a): Vertical Re-Entrant channel configuration 33 Figure 3.2 (b): Vertical Re-Entrant channel configuration 33 Figure 3.3: Proposed New Fuel Channel 34 vii

8 Figure 3.4 (a): Fuel length of Re-Entrant channel 35 Figure 3.4 (b): Entrance Region of Re-Entrant channel 35 Figure 3.4 (c): Re-Entrant region of Re-Entrant channel 35 Figure 3.5: Cross-sectional view of Re-Entrant channel 40 Figure 4.1: Re-Entrant flow-channel showing nodes and initial temperature estimates Figure 4.2: Thermal Conductivity of porous YSZ Figure 4.3: Flow chart representing Numerical Model in MATLAB 56 Figure 5.1: Axial temperature profile along the fuelled channel length for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.2: Specific heat, thermal conductivity and Prandtl number profiles along the fuel-channel for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.3: Average specific heat, average Prandtl number, Heat Flux and HTC profiles along the fuel-channel for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.4: Outer-sheath-temperature profile along the channel length for powers of 8.5 MW th and 5.5 MW th, inner-tube thickness of 2 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.5: Outer-sheath temperature profiles with variable flow-tube thickness, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.6: Change in temperature across the flow tube for different wall thermal conductivities, inner-tube thickness of 2 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.7: Temperature Profile of Outer Sheath Temperature for variable wall thermal conductivity, inner-tube thickness of 2 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux viii

9 Figure 5.8: Axial temperature profile along the fuelled channel length for a channel power of 10 MW th, inner-tube thickness of 2 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.9: Temperature profile along the channel length for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.10: Specific heat, thermal conductivity and Prandtl number profiles along the fuel-channel for a channel power of 8.5 MW, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.11: Average specific heat, average Prandtl number and HTC profiles along the fuel-channel for a channel power of 8.5 MW, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.12: Temperature gradients along the radial distance from center for a non insulated Re-Entrant channel Figure 5.13: Heat loss to the moderator when pressure tube thickness is varied for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.14: Change in temperature across the pressure tube when thermal conductivity is varied for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.15: Temperature profile along the Porous YSZ insulated channel length for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.16: Temperature gradients along the radial distance from center for a Porous YSZ insulated Re-Entrant channel for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, insulation thickness of 7 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.17: Temperature profile along the ZrO 2 insulated channel length for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, insulation thickness of 7 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux ix

10 Figure 5.18: Temperature gradients along the radial distance from center for a ZrO 2 insulated Re-Entrant channel for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, insulation thickness of 7 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure 5.19: Heat loss to Moderator for cold side of Re-Entrant channel for variable Porous YSZ insulation thickness and variable solid ZrO 2 insulation thickness assuming mixed type heat loss to the moderator Figure 5.20: Heat loss to Moderator for cold side of Re-Entrant channel for variable Porous YSZ insulation thickness and variable solid ZrO 2 insulation thickness assuming free convection heat loss to the moderator Figure 5.21: Axial Power Profiles 95 Figure 5.22: Case 1 (Uniform Power Profile): Temperature profiles along the 7 mm thick Porous YSZ insulated fuel-channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure 5.23: Case 2 (Typical Nominal Power Profile): Temperature profiles along the 7 mm thick Porous YSZ insulated fuel-channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure 5.24: Case 3 (Peaked Shape Power Profile): Temperature profiles along the 7 mm thick Porous YSZ insulated fuel-channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure 5.25: Case 4 (Upstream-skewed Power Profile): Temperature profiles along the 7 mm thick Porous YSZ insulated fuel-channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure 5.26: Case 5 (Downstream-skewed Power Profile): Temperature profiles along the 7 mm thick Porous YSZ insulated fuel-channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure 5.27: Case 6 (Variable Power Profile): Temperature profiles along the 7 mm thick Porous YSZ insulated fuel-channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s x

11 Figure B.1: Comparison of coolant temperature for a total channel power of 8.5 MW th, mass flow rate of 4.37 kg/s, uniform heat flux and no heat loss to the moderator Figure C 1(a): Specific heat, thermal conductivity and Prandtl number profiles along a YSZ insulated fuel-channel for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure C 1(b): Average specific heat, average Prandtl number and HTC profiles along a YSZ insulated fuel-channel for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure C 2 (a): Specific heat, thermal conductivity and Prandtl number profiles along a ZrO 2 insulated channel for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, insulation thickness of 7 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure C 2(b): Average specific heat, average Prandtl number and HTC profiles along a ZrO 2 insulated channel for a channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, insulation thickness of 7 mm, inlet temperature of 350 C, mass flow rate of 4.37 kg/s and uniform heat flux Figure D.1 (a): Case 2 - Specific heat, Prandtl number and thermal conductivity profiles along a 7 mm thick Porous YSZ insulated channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure D.1 (b): Case 2 - Average specific heat, average Prandtl number and HTC profiles along a 7 mm thick Porous YSZ insulated channel for a total channel power of 8.5 MW, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure D.2 (a): Case 3 - Specific heat, Prandtl number and thermal conductivity profiles along a 7 mm thick Porous YSZ insulated channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure D.2 (b): Case 3 - Average specific heat, average Prandtl number and HTC profiles along a 7 mm thick Porous YSZ insulated channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s xi

12 Figure D.3 (a): Case 4 - Specific heat, Prandtl number and thermal conductivity profiles along a 7 mm thick Porous YSZ insulated channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure D.3 (b): Case 4 - Average specific heat, average Prandtl number and HTC profiles along a 7 mm thick Porous YSZ insulated channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure D.4 (a): Case 5 - Specific heat, Prandtl number and thermal conductivity profiles along a 7 mm thick Porous YSZ insulated channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure D.4 (b): Case 5 - Average specific heat, average Prandtl number and HTC profiles along a 7 mm thick Porous YSZ insulated channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure D.5 (a): Case 6 - Specific heat, Prandtl number and thermal conductivity profiles along a 7 mm thick Porous YSZ insulated channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s Figure D.5 (b): Case 6 - Average specific heat, average Prandtl number and HTC profiles along a 7 mm thick Porous YSZ insulated channel for a total channel power of 8.5 MW th, inner-tube thickness of 2 mm, pressure tube thickness of 11 mm, inlet temperature of 350 C and mass flow rate of 4.37 kg/s xii

13 NOMENCLATURE h R Q area, average specific heat, diameter, gravity, mass flux, enthalpy, heat transfer coefficient, thermal conductivity, length, mass-flow rate, perimeter, thermal resistance, heat transfer rate, total heat transfer rate, temperature,, fluid velocity in x-direction, fluid velocity in y-direction, axial location, xiii

14 Greek Letters α β ν thermal diffusivity, expansion coefficient, thickness, dynamic viscosity, viscosity, density, Dimensionless numbers Nusselt number Prandtl number average Prandtl number Ra Rayleigh number Reynolds number Subscripts and Superscripts A b annulus bulk xiv

15 B c CE cr cs E f forced free fs FT hs hy i IT max mod n o os p pc PT r total bundles cross section centre element critical cold side fuel element flow forced convection free convection fuel sheath flow tube hot side hydraulic inner inner tube maximum moderator node outer outer sheath perimeter pseudocritical pressure tube radius total (free and forced) convection xv

16 w wet x wall wetted axial length along fuel-channel Acronyms ACR CANDU FT GFR GIF IT HEC HTC LFR MATLAB MSR Mtoe NIST OT PT PV REC REFPROP SC Advanced CANDU Reactor CANada Deuterium Uranium Flow Tube Gas-cooled Fast Reactor Generation IV International Forum Inner Tube High Efficiency Channel Heat Transfer Coefficient Lead-cooled Fast Reactor MATrix LABoratory Molten Salt Reactor Million tonnes of oil equivalent National Institute of Standards and Technology Outer Tube Pressure Tube Pressure Vessel Re-Entrant Channel REFerence fluid thermodynamic and transport PROPerties SuperCritical xvi

17 SCW SCWR SFR SuperCritical Water SuperCritical Water-cooled Reactor Sodium-cooled Fast Reactor SS-304 Stainless Steel 304 VHTR YSZ Very-High Temperature Reactor Yttria Stabilized Zirconia xvii

18 CHAPTER 1 INTRODUCTION Energy sources available today include fossil fuels, nuclear, hydroelectric, gas, wind, solar, refuse-based and biomass technologies. While the overall energy demand in developed countries has levelled-out in recent decades, the growth in energy demand of developing countries is increasing the pressure on energy resources worldwide. This growth is projected to increase as the world s population grows from the present level of 6 billion to 8 billion in 2025, and as the people in developing nations aspire to have a higher standard of living. Energy demand is derived from three major sectors, namely, domestic and commercial, industry and agriculture, and transport. There is a significant demand for energy in the form of electricity for the first two sectors. Electricity generation presently uses about 40% of the world s energy supply [1]. The future total energy growth worldwide is projected to average 1.7% per year to The demand is expected to reach 191,895 TWh in 2030 as compared to 119,789 TWh in While the demand in 2002 corresponds to 10,300 million tonnes of oil equivalent (Mtoe), the future demand corresponds to 16,500 Mtoe, where Mtoe is the amount of energy released by burning one million tonnes of crude oil. World electricity demand is projected to rise from 16,000 TWh/year in 2002 to 31,600 TWh/year in 2030, at an average rate of 2.5% per year [2]. Thus, electricity demand is growing much faster than overall energy demand. While demand for oil continues to rise, available resources are declining and consumption is twice the rate of discovery of new oil resources. Most natural gas reserves are located in geopolitically uncertain areas and, hence transport becomes a major problem. Furthermore, natural gas production is likely to approach its peak in the next couple of decades in many countries. Renewable sources of energy cannot meet the extent of the future demand as the intermittent nature of these sources cannot be controlled to provide either continuous base-load power or peak-load power when it is needed. Coal resources are abundant worldwide and coal is economically attractive to 1

19 use in large scale, but it is a significant contributor to greenhouse gas emissions among all fossil fuels and has significant health related effects[1]. Nuclear power on the other hand is comparatively cleaner. Presently, there are about 440 nuclear power reactors operating in 30 countries worldwide, producing 16% of the world s electricity. 15% of Canada s electricity is generated from nuclear power while over 50% of the electricity generated in the province of Ontario is from nuclear power [3]. Assuming the current market share is maintained, the electricity growth translates to 2496 TWh of nuclear power over 28 years. This corresponds to an increase of 89 TWh of nuclear power each year. As traditional fossil fuel power plants are phased out to reduce greenhouse gas emissions, nuclear energy is the only non-greenhouse gas emitting power source that can effectively replace the fossil fuel generated electrical power in sufficient quantity and satisfy increasing global demand. First generation nuclear reactors were built in the 1950s in Canada. Generation II and Generation III CANada Deuterium Uranium (CANDU) nuclear reactors are currently being used for energy production. In the CANDU design, fission reactions heat the heavy water coolant, which is heavy water in current CANDU reactors, to produce steam which is then used to generate electricity from a turbine generator system. Generation III+ CANDU reactors (e.g. ACR-1000) are currently being developed which would use light-water as the coolant. Research is looking ahead into the next generation, or Generation IV nuclear reactors, which are expected to be more efficient and cheaper to build than current reactor designs An international effort established the Generation IV International Forum (GIF) in 2001 to indentify and select nuclear energy systems for further development. Argentina, Brazil, Canada, France, Japan, the Republic of Korea, the Republic of South Africa, the United Kingdom, and the United States signed the GIF charter in 2001 while Switzerland, Euratom, the People s Republic of China, and the Russian Federation signed the charter later on. Six different reactor systems were selected, namely, Gas-cooled Fast Reactors (GFRs), Lead-cooled Fast Reactors (LFRs), Molten Salt-cooled Reactors (MSRs), Sodium-cooled Fast Reactor (SFRs), SuperCritical Water-cooled Reactors (SCWRs), and 2

20 Very High-Temperature gas-cooled Reactors (VHTRs). Of these types, the SCWR is Canada s premier choice for Generation IV Reactor technology with some materials related research on VHTRs. SuperCritical Water-cooled nuclear Reactors use light water as the coolant and operate at supercritical pressures and temperatures, that is, pressures and temperatures above the critical point of water. The critical pressure of water is MPa and the critical temperature is C [4]. At the end of the 1950s and 1960s, research was conducted to investigate the possibility of using supercritical fluids in nuclear reactors and several nuclear reactor design concepts using supercritical water were developed in the United States and the former USSR. This idea was abandoned, likely due to material constraints [5]. SCWRs have re-emerged as a viable option for Generation IV nuclear reactors now that there is more experience with fossil-based SuperCritical Water (SCW) plants and advanced materials have been developed for use in SCW environments. The main advantages of SCWRs are: 1) an increase in thermal efficiency of nuclear power plants from % to % which corresponds to current thermal (fossil fuel type) power plants; 2) an expected decrease in capital and operational costs, hence reduction in electrical-energy costs; 3) a simplified flow circuit with the elimination of steam dryers, steam heat separators etc. (for direct cycle version); and 4) the ability to facilitate steam based technologies such as desalination, thermochemical hydrogen production, or district heating due to higher temperature. There are two types of SCWRs currently being developed: (i) Pressure Vessel type SCWR, and (ii) Pressure Tube type SCWR. The latter has been chosen by Canada and Russia as their SCWR design concept. While current CANDU-6 reactors operate at a coolant pressure range of MPa, and current PWRs operate at a coolant pressure range of MPa, SCWRs will operate at about 25 MPa. The inlet and outlet design 3

21 temperatures for the CANDU-SCWR is expected to be 350 C and 625 C respectively [6]. Supercritical water has unique properties such as a liquid-like density and a gas like viscosity [7]. There are significant changes in thermophysical properties such as specific heat, density, viscosity, and thermal conductivity of supercritical water within ±25 C from the pseudocritical temperature [8]. The density, viscosity, and thermal conductivity drastically decrease with increasing temperature through the pseudocritical region while the specific heat increases and then decreases in this region. The current CANDU fuel-channel design cannot be used for the SCWR as the higher pressures in the SCWR fuel-channel will cause the pressure tube to rupture. Thus, a new fuel-channel design needs to be developed. The current SCWR channel-type design concept uses a ceramic liner, to reduce heat losses to the moderator, with a perforated metal insert to protect the ceramic liner from fuel bundles [9]. While such a design may work, alternative design concepts are under development to explore the optimum efficiency of a SCWR channel-type reactor. An alternative design being considered is the Re-Entrant fuel-channel (REC) which consists of three tubes: the inner tube (flow tube), the pressure tube, and an outer tube. The fuel bundles, similar to those of the current CANDU reactors, are placed in the inner tube. The flow and pressure tubes form an annulus through which flows the primary coolant. At the far end, the flow will reverse direction and enter the inner tube, and hence the fuel string. The coolant exits the channel from the inner tube. Insulation options such as ceramic and carbon dioxide gas can be used in the gap between the pressure tube and the outer tube. As the coolant enters the channel below the pseudocritical temperature and exits the channel well above the pseudocritical temperature, it passes through the pseudocritical region along the heated length of the fuel-channel [5]. The location of the pseudocritical point is one of the main reasons for developing the REC. As deteriorated heat transfer occurs around the pseudocritical point and as there 4

22 are also significant material chemistry issues in the pseudocritical region, the occurrence of the pseudocritical point in the annulus of the REC, if possible, may address some of these concerns. Another reason for developing the REC is that the coolant can act as a preheater in the annulus and can reduce heat losses. There are also concerns with the use of the ceramic insulator and perforated liner as unproven technology in supercritical water. An advantage of having the ceramic insulator outside the pressure tube in the REC design is that the ceramic will never come in contact with the coolant, hence avoiding the problem of creating ceramic particles that could enter the primary heat transport system. An advantage of using carbon dioxide gas in the ceramic insulating region in the REC is that it helps in detecting a leak in the pressure tube by analyzing the moisture content in the gas. A disadvantage of the REC is that it may lead to an increased complexity in the end fitting design [9]. The pseudocritical region also affects various thermophysical properties of the coolant which impacts the temperature profiles of the coolant and of the outer sheath. The pseudocritical point can either occur in the inner tube or in the annulus of the REC. Thus, it is important to know the location of the pseudocritical point in the channel. This can be done by developing a numerical model of the REC and by modelling the heat transfer across the fuel-channel. 1.1 Objective Therefore, the primary objective of this thesis is to develop a design concept of an alternative SCWR Fuel Channel concept called the Re- Entrant Fuel Channel and to assess the thermal hydraulic behaviour of the channel. Additional objectives of the work are as follows: 1. Calculate temperature profiles of the coolant, flow tube, pressure tube, outer tube and outer sheath along the length of the fuel-channel to determine the effect of the pseudocritical point on the temperature profiles, and to verify that the outer sheath temperature is below the sheath temperature limit; 5

23 2. Calculate the total heat loss from the Re-Entrant Fuel Channel to the moderator and to evaluate the efficiency of the REC; 3. Optimize the design of the channel by performing various sensitivity analyses; and to 4. Compare the total heat loss and efficiency with the existing SCWR channel-type design concept. Chapter 2 of this thesis will describe the literature review with particular focus on SCWR type reactors, supercritical fluid properties, and the general thermal hydraulic behaviour of supercritical water. Chapter 3 will discuss the methodology and the proposed design options for the REC. The numerical model of the REC is described in Chapter 4. Using the numerical model, a heat transfer analysis was conducted considering no heat loss to the moderator, heat losses to the moderator with different insulators, the impact of non-uniform heat flux shapes, and the impact of boundary conditions. The results of the heat transfer analysis are discussed in detail in Chapter 5. The design of the REC can thus be optimized considering variable power, flow, and initial conditions. Concluding remarks and future work are described in Chapters 6 and 7. 6

24 CHAPTER 2 BACKGROUND AND LITERATURE REVIEW 2.1 Generation IV Nuclear Technology Figure 2.1 shows the advancement of nuclear reactor technology in Canada since the 1950s, when nuclear power was first used for commercial production of electricity.. Nuclear Power Demonstration (NPD), a small scale prototype CANDU type reactor and Douglas Point, a larger prototype, commenced operation in 1962 and 1967 respectively. These reactors are referred to Generation I nuclear reactors and they established the technological base necessary for larger commercial CANDU units. CANDU 6 reactors are Generation II reactors. The first commercial CANDU unit came into operation in 1971 in Pickering, Ontario [10]. CANDU type reactors use heavy water (D 2 O) as their coolant and moderator, operate at a pressure of approximately 10 MPa and have inlet and outlet temperatures of approximately 260 C and 310 C respectively [11]. The CANDU 3 and CANDU 9 were Generation III reactor concepts which were designed, but never built. The Enhanced CANDU 6 (EC6) is part of the Generation III reactors which is currently being developed by AECL. It retains the basic features of the CANDU 6 design, but incorporates newer technologies that enhance safety, operation, and performance [12]. The Advanced CANDU reactor (ACR) is a Generation III+ reactor that is also currently being developed by AECL. The ACR is a light-water cooled and heavy-water moderated reactor [13,14]. 7

25 8 Figure 2.1: Nuclear Reactor Technology in Canada [12,13,,14]

26 The Generation IV International Forum (GIF) was established in 2001 to select and develop the next generation of nuclear energy systems known as Generation IV. Canada, Argentina, Brazil, France, Japan, the Republic of Korea, the Republic of South Africa, the United Kingdom, and the United States signed the GIF in Switzerland joined the forum in 2002, Euratom in 2003 and Russia joined in 2006 along with China. The main requirements for Generation IV reactors as established by the GIF are enhancements in sustainability, safety and reliability, economical viability, proliferation resistance and physical protection [15]. These requirements are listed below: Sustainability The Generation IV nuclear energy systems will provide sustainable energy generation that meets clean air objectives and provides longterm availability of systems and effective fuel utilization for worldwide energy production. The Generation IV nuclear energy systems will minimize and manage their nuclear waste and notably reduce the long-tem stewardship burden, thereby improving protection for the public health and environment. Economics Generation IV nuclear energy systems will have a clear life-cycle cost advantage over other energy sources and will have a level of financial risk comparable to other energy projects. Safety and Reliability Generation IV nuclear energy systems operations will excel in safety and reliability. These systems will have a very low likelihood and degree of reactor-core damage and will eliminate the need for offsite emergency response. Proliferation Resistance and Physical Protection Generation IV nuclear energy systems will offer enhanced security for nuclear materials and facilities against acts of terrorism. 9

27 Generation IV nuclear energy systems are intended to meet the above requirements in order to provide a sustainable development of nuclear energy. 2.2 Generation IV Concepts The GIF panel selected six reactor systems for further development from 100 concepts. A brief description of these systems is shown below: Very-High Temperature Reactors (VHTRs) The VHTR is a helium-gas cooled, graphite moderated and thermal neutron spectrum reactor. The reference parameters are a coolant inlet and outlet temperatures of approximately 500 C and 1000 C respectively, an operating pressure of 9 MPa and reactor thermal power of 600 MW. The VHTR will be mainly used for the cogeneration of electricity and hydrogen as well as other process heat applications [16]. Sodium-cooled Fast Reactors (SFR) Liquid sodium is used as the reactor coolant in this type of fast neutron spectrum reactor whose outlet temperatures ranges from 530 C to 500 C. The SFR is designed for efficient management of high-level wastes such as plutonium and other actinides. Three reactor sizes are being designed, large (600 to 1500 MW) loop-type reactor, intermediate (300 to 600 MW) pool-type reactor and small (50 to 150 MW) modular-type reactor [16,17]. Gas-cooled Fast Reactors (GFRs) GFRs use a direct Brayton cycle gas turbine for electricity and hydrogen production with high efficiency. The 1200 MW reference reactor has a fast neutron spectrum, an inlet temperature of 485 C, outlet temperature of 850 C and an operating pressure of 7 MPa [16,18]. Lead-cooled Fast Reactors (LFRs) LFRs use molten lead or lead-bismuth as an inert coolant and have a fast neutron spectrum. The main purpose of LFRs is 10

28 electricity and hydrogen production along with actinide management. LFRs have an inlet temperature of C and an outlet temperature of C [19]. Molten Salt Reactors (MSRs) MSRs have a thermal neutron spectrum and use molten salts such as sodium fluoride salt as their coolant. The coolant outlet temperature of MSRs can range up to 800 C and the reference thermal power is 1,000 MW [16,20] SuperCritical Water-cooled Reactors SCWRs use Super-critical water as their coolant and operate above the thermodynamic critical point of water to increase efficiency. Both thermal neutron and fast neutron spectra are being considered for this type of reactor. SCWRs operate at 25 MPa and have an outlet temperature of up to 625 C. There are two different types of SCWRs, Pressure Vessel (PV) type SCWR and Pressure Tube (PT) type SCWR [16,21]. VHTRs and SCWRs are Canada s choice for Generation IV reactors. Extensive research is currently being conducted to develop these two reactor concepts SuperCritical Water-cooled Reactor Concepts As previously mentioned, there are two different types of SCWRs that are currently being developed worldwide. The PV type reactor as seen in Figure 2.2, is a large reactor pressure vessel which has a wall thickness of 0.5 m to contain the reactor core. This type of reactor is analogous to conventional Light Water Reactors and is being developed by the United States. The PT reactor, which can be seen in Figure 2.3, is analogous to conventional Heavy Water Reactors and is designed to be more flexible with respect to flow, flux and density changes compared to PV reactors. The PT reactor concept is being developed in Canada 11

29 and Russia by Atomic Energy of Canada Limited (AECL) and the Research and Development Institute of Power Engineering (RDIPE and NIKIET) respectively [8]. The separation between the moderator and the coolant in the PT SCWR concept allows for significant enhancement in safety. While SCW is the coolant, various moderator options are currently being considered. Three thermodynamic cycle options are being considered for SCWRs. They include the direct, indirect, and dual cycle options. SC steam from the reactor is directly fed to a SC turbine in the direct cycle. This eliminates the need for steam generators which results in the highest cycle efficiency among the three cycles. The indirect and dual cycles use heat exchangers to transfer heat from the reactor to the turbine. Two types of heat exchangers are being considered: SC water to SC water and SC water to superheated steam. The maximum temperature of the secondary loop is lower due to heat transfer through the heat exchangers, hence lowering the efficiency of the cycle [22]. The above mentioned three options are further divided into various re-heat options. A direct cycle SCWR with no-re-heat option has been chosen for this analysis and the T-s diagram is shown in Figure 2.4. Saturated water enters the pump (Point 1), which compresses the fluid to the supercritical operating pressure. The water is then pre-heated (Point 2 Point 3), before it enters the fuel-channel. The coolant continues to be heated as it flows through the fuel-channel (Point 3 Point 4). The SC steam outlet from the reactor, which is at pressure of 25 MPa and temperature of 625 C, is then expanded in the SCW turbine (Point 4 Point 5) to a sub-atmospheric pressure of 6.77 kpa. The steam is condensed inside a condenser and enters the pump (Point 5 Point 1), thus completing the thermodynamic cycle [23]. 12

30 25 MPa (supercritical) 500 C (supercritical) 25 MPa (supercritical) 280 C (subcritical) Subcritical pressure Subcritical temperature Figure 2.2: Pressure Vessel Type SCWR [5] Sustainable Fuel input Multiple products are key to sustainable future and competitive designs Electric power Hydrogen and process heat Core Pump Turbine Generator H.P Heat for Co- Generation or IP/LP Turbines Drinking water T3, P3 Turbine H.P. S T2, P2 CONDENSER T1, P1 Brine Industrial isotopes Figure 2.3: Pressure Tube Type SCWR [5] 13

31 Figure 2.4: T-s diagram for direct cycle SCWR with no-reheat option [23] 14

32 The main advantage of SCWRs is an increase in thermal efficiency of Nuclear Power Plants from % to % which corresponds to current thermal power plants. SCWRs operate on a direct cycle where the coolant is the reactor is used in the turbines, thus eliminating the need for steam dryers, steam heat separators etc., which would lead to a simplified flow circuit and lower capital and operational costs [5]. SCWRs also have the ability to facilitate steam based technologies such as desalination, thermochemical hydrogen production and district heating [24]. Supercritical water has unique thermophysical properties and these properties are discussed in the following section. 2.3 SCW Properties The thermophysical properties of supercritical water undergo significant changes within the critical and pseudocritical regions. The critical point occurs when the distinction between the liquid and vapour phases disappears and is characterized by T cr and P cr. The critical pressure for water is MPa and the critical temperature of water is C [4]. The pseudocritical point is a point at a pressure above the critical pressure and at a temperature corresponding to the maximum specific heat for this particular pressure [5]. Since SCWRs will operate at a pressure of 25 MPa, the pseudocritical temperature can be identified by the peak in the specific heat curve shown in Figure 2.5 (a). From Figure 2.5 (a), the pseudocritical temperature corresponds to C. Thermophysical properties such as density, thermal conductivity, dynamic viscosity, kinematic viscosity, and Prandtl number undergo significant changes in the pseudocritical region. These changes can be seen in Figure 2.5 (b-f) (Figures constructed using values extracted from NIST REFPROP at 1 C intervals). 15

33 80 70 pc point P = 25 MPa 60 Specific Heat [kj/kgk] Temperature [ C] Figure 2.5 (a): Specific Heat vs. Temperature of water in the pseudocritical region at 25 MPa 16

34 pc point P = 25 MPa 500 Density [kg/m 3 ] Temperature [ C] Figure 2.5 (b): Density vs. Temperature of water in the pseudocritical region at 25 MPa 17

35 0.6 pc point P = 25 MPa 0.5 Thermal Conductivity, k [W/mK] Temperature [ C] Figure 2.5 (c): Thermal Conductivity vs. Temperature of water in the pseudocritical region at 25 MPa 18

36 10 9 pc point P = 25 MPa 8 Prandtl Number (based on T b ) Temperature [ C] Figure 2.5 (d): Prandtl Number vs. Temperature of water in the pseudocritical region at 25 MPa 19

37 80 70 pc point P = 25 MPa 60 Dynamic Viscosity [µpas] Temperature [ C] Figure 2.5 (e): Dynamic Viscosity vs. Temperature of water in the pseudocritical region at 25 MPa 20

38 pc point P = 25 MPa Kinematic Viscosity [cm 2 /s] Temperature [ C] Figure 2.5 (f): Kinematic Viscosity vs. Temperature of water in the pseudocritical region at 25 MPa 21

39 The most significant changes occur within ±25 C of the pseudocritical temperature. Properties such as density and dynamic viscosity undergo a significant drop in this region while there in an increase in the kinematic viscosity. Thermal conductivity and Prandtl number peak at the pseudocritical point, albeit the peak in the thermal conductivity is relatively small. As the inlet temperature for a SCWR is 350 C and the outlet temperature is 625 C, the coolant would pass through the pseudocritical region before reaching the channel outlet [5] Various correlations have been developed to calculate the heat transfer characteristics of supercritical water as shown in the following section. 2.4 SCW Correlations Dyadyakin and Popov developed a correlation for supercritical water heat transfer for fuel bundles [25].. = µ. µ (2.1) where is the axial location along the heated length in metres and is the hydraulic diameter. A tight-lattice, 7 element, helically finned, water-cooled bundle cooled with water was used to develop the correlation. As heat transfer correlations for bundles are generally very sensitive to bundle design because of the effect of different bundle components, and the experiments appear to be for a mobile-type reactor, this correlation cannot be applied to SCWRs [8]. 22

40 The Dittus-Boelter correlation shown in Equation (2.2) is the most widely used heat transfer correlation at subcritical pressures for forced convection. The use of this correlation as shown in Equation (2.2) was proposed by McAdams for forced convective heat transfer in turbulent flows at subcritical pressures [26]. = (2.2) Equation (2.2) was later used at supercritical conditions. According to Schnurr et al., Equation (2.2) showed good agreement with experimental data for supercritical water flowing inside circular tubes at a pressure of 31 MPa. It was also noted that the equation might produce unrealistic results within some flow conditions near the critical and pseudocritical points but this equation was used as a base for other supercritical heat transfer correlations [27]. The original Dittus-Boelter correlation shown above is used in the following form for reference purposes [28]: = (2.3) Bishop et al. conducted experiments with supercritical water flowing upward inside tubes and annuli for the following range of operating parameters: pressure: MPa, bulk fluid temperature: C, mass flux: kg/m 2 s and heat flux : MW/m 2. Their data for heat transfer were generalized using the following correlation:. = (2.4) 23

41 where is the axial location along the heated length, is the density of the fluid at the wall temperature, is the density of the fluid at bulk temperature and the last term accounts for entrance-region effects [29]. The Bishop et al. correlation is often used without the entrance-region term and is hence written as:. = (2.5) Swenson et al. investigated heat transfer coefficients in smooth tubes and found that conventional correlations do not work well as they use bulk fluid temperature to calculate thermophysical properties. The Swenson et al. correlation was developed using a pressure range of MPa, bulk fluid temperature of C and mass flux of kg/m 3 s. The correlation uses the wall temperature to calculate thermophysical properties and is shown in Equation (2.6) [30].. = (2.6) Mokry et al. modified the existing correlations for SCW to obtain a new correlation for forced-convective heat transfer in a vertical bare tube as seen in Equation (2.7) [8]:. = (2.7) Zahlan et al. compared many empirical correlations using a dataset provided by Kirillov et al. from the Institute of Physics and Power Engineering (Obninsk, Russia) and concluded that the Mokry et al. correlation is currently the best known correlation for heat transfer in SCW [31,32]. 24

42 A comparison of the previously mentioned correlations is shown in Figure 2.6. The bulk fluid temperature was assumed to be 350 C and the wall temperature was assumed to be 400 C. From the figure, it can be seen that the Mokry et al. correlation is the most conservative of the four correlations shown. Thus, the Mokry et al. correlation will be used for a conservative analysis in this work. The heat transfer calculations should be modified as new heat transfer correlations for SCW are published. 25

43 10000 Bishop et al. corr Dittus-Boelter corr. log Nu Swenson et al. corr. 100 Mokry et al. corr log Re Figure 2.6: Comparison between SCW correlations when bulk fluid temperature is 350 C and wall temperature is 400 C 26

44 2.5 Fuel Channel Design Concepts The current CANDU fuel-channel found in operating CANDU reactors consists of two tubes, the pressure tube and the calandria tube. The fuel bundles are located within the pressure tube. An annulus gas gap, which provides thermal insulation, separates the calandria tube and the pressure tube. The moderator flows outside the calandria tube. The pressure tube is made from a zirconium niobium alloy while the calandria tube is made from zirconium [11]. Figure 2.7 shows the present CANDU type fuel-channel. CANDU 6 reactors have an operating pressure of MPa and inlet and outlet temperatures of 260 C and 310 C respectively while SCWRs will have an operating pressure of 25 MPa and inlet and outlet temperatures of 350 C and 625 C [11]. The higher pressures associated with SCWRs are above the pressure tube burst pressure and thus the current design must be changed. The current candidate for the CANDU SCWR PT type fuel-channel design uses only a pressure tube as shown in Figure 2.8. A ceramic liner is used to reduce heat losses to the moderator. A perforated metal insert is used to protect the ceramic liner from fuel bundles from scratching during refuelling and also reduce the erosion of the ceramic liner. The premise for this design is that the metal insert acts as a fuelling sleeve for transfer of the fuel while the ceramic acts as a thermal barrier. In doing so, the pressure tube will be at the moderator temperature and the thermal component of pressure tube creep will be significantly reduced. This design is also called the High Efficiency Channel (HEC) [9]. 27

45 Figure 2.7: CANDU type fuel-channel Figure 2.8: SCWR type fuel-channel concept 28

46 While such a design may work, there are concerns with the construction, assembly, and maintenance of the HEC. One potential problem with this design is that if the ceramic insulator erodes, fractures, or chips, it might be difficult to repair or replace the ceramic insulator due to the presence of the metal insert. If deterioration of the ceramic liner was to occur, then the thermal barrier is weakened and potential hotspots could occur on the pressure tube. This may result in some of the current problems such as hydriding and blistering occurring in HEC design. Hence, alternative design concepts that will not have the above mentioned problem are being considered for the SCWR channel-type reactor. One such alternative design being considered is called the Re-Entrant Fuel Channel and is explained in the following Chapter. 29

47 CHAPTER 3 PROPOSED FUEL CHANNEL DESIGN CONCEPT The design requirements for the proposed fuel-channel concept are as follows: 1. The fuel-channel should be able to be used in a horizontal or vertical configuration. 2. The fuel-channel components should be able to withstand the high operating pressures and temperatures. 3. The fuel-channel components should be as neutron transparent as possible. 4. The fuel-channel should be insulated if required, to reduce heat losses to the moderator. The proposed fuel-channel design concept under consideration consists of three tubes; an inner tube (flow tube), a pressure tube, and an outer tube. The fuel bundles are located within the inner tube, while a ceramic insulator or other thermal resistance barrier is placed in-between the pressure tube and the outer tube to reduce heat losses from the fuel-channel to the moderator. The outer tube protects the ceramic insulator from the heavy water moderator system. Figure 3.1 (a) shows the proposed new fuel-channel concept in a horizontal configuration in the calandria layout. The coolant inlet and outlet are in the same side of the new fuel-channel. The end fittings support the fuel-channels and allow a pressure tight connection with the fuelling machines, while a removable closure plug closes each end fitting. The end shields are filled with light water and steel balls, which thermalize and absorb neutrons. Figure 3.1 (b) shows a possible horizontal channel layout of 300 fuel-channels in the calandria vessel. The vertical channel configuration of the proposed new fuel-channel is similar to the horizontal configuration except for the calandria vessel, which will be closer to a pressure vessel in this configuration. The vertical channel configuration is shown in Figure 3.2 (a), while a possible vertical channel layout is shown in Figure 3.2 (b). 30

48 The coolant first flows through the gap between the pressure tube and the flow tube from one end of the channel to the other before reversing direction and flowing through the inner tube. Thus, the fuel-channel effectively becomes a double-pipe heat exchanger in which the annulus acts as a preheater. The Re-Entrant channel design is shown in Figure 3.3. The fuel length of the current CANDU-type fuel-channel, m, is chosen as the reference fuel length of the new fuel-channel. The side view of the Re-Entrant channel is shown in Figure 3.4 (a). The inner tube is referred to as the hot side and the annulus is referred to as the cold side of the Re-Entrant fuel-channel. Figures 3.4 (b) and (c) show the entrance region and the re-entrant region of the fuel-channel respectively. The mass flow rate of the coolant in each channel is 4.37 kg/s [24]. The moderator temperature is estimated to be 80 C based on current operating parameters. The mean mass-flow rate of 0.95 kg/s for a mixed type flow in a CANDU-6 reactor is chosen as the reference mass flow rate of the moderator outside the Re-Entrant channel [33]. The reference case is a Channel Thermal Power of 8.5 MW th uniformly applied in the fuelled region. Variable power profiles have also been accounted for in this work. 31

49 Figure 3.1 (a): Horizontal Re-Entrant channel configuration Figure 3.1 (b): Horizontal Re-Entrant channel configuration 32

50 Figure 3.2 (a): Vertical Re-Entrant channel configuration Figure 3.2 (b): Vertical Re-Entrant channel configuration 33

51 34 Figure 3.3: Proposed New Fuel Channel

52 Figure 3.4 (a): Fuel length of Re-Entrant channel Figure 3.4 (b): Entrance Region of Re-Entrant channel Figure 3.4 (c): Re-Entrant region of Re-Entrant channel 35

53 The inner diameter of the flow tube is kept equal to that of the current CANDU-type fuelchannel for two reasons. The first is that mm allows current CANDU-type fuels to be used in the Re-Entrant channel. The second is that the industry has a lot of manufacturing and operating experience with this diameter range. The flow tube inner diameter is a design parameter that could be changed once more knowledge on the reactor physics and recommended fuel types is known. The four different options for the fuel bundles are shown in Table 3.1 [34]. They are the 37-Element bundle, the CANFLEX bundle, the Variant-18 bundle, and the Variant-20 bundle. The 37-Element and the CANFLEX bundles are currently being used, while the Variant-18 and the Variant-20 bundle concepts are variations of the CANFLEX bundle that are presently being developed. The CANFLEX, Variant-18, and the Variant-20 bundles have a total of 43 elements, while the 37-Element bundle has 37 elements. All four fuel bundles contain four rings of elements and the number of elements in each ring is different for the 37- Element bundle and the 43 element bundles, as seen in Table 3.1. Unlike the 37-Element bundle and the CANFLEX bundle, the centre element in the Variant-18 and Variant-20 bundles is unheated and filled with Dysprosium, a burnable neutron absorber which can reduce void reactivity [35]. The heated length of all four bundles is 481 mm, which corresponds to a total of 12 bundles in the m long Re-Entrant fuel-channel. For the purposes of this work, the Variant-18 fuel bundle concept has been chosen for the horizontal Re-Entrant channel configuration. Note that the vertical configuration will use a fuel string and, depending on the fuelling method, a central rod might be needed to connect to the fuel string. 36

54 Table 3.1: Parameters of current fuel bundle options for the Re-Entrant channel Parameter Value 37 Element CANFLEX Variant-18 Variant-20 Total No. of Elements No. of elements in the centre ring No of elements in the inner ring No. of elements in the intermediate ring No of elements in the outer ring Outer diameter of centre ring element (mm) Outer diameter of inner ring element (mm) Outer diameter of intermediate ring element (mm) Outer diameter of outer ring element (mm) Heated Bundle Length (mm)

55 There are few options for the material of construction of the tubes in the Re-Entrant channel. They include, Zirconium alloy with 2.5 wt% Niobium (Zr-2.5Nb) which is presently used as the material of construction of CANDU-type pressure tubes, Stainless Steel Grade 304 (SS-304), Inconel-718, and a zirconium alloy called Excel (Zr-3.5% Sn-0.8% Nb 0.8% Mo-1130 ppm O). The Ultimate Tensile Strength (UTS) for all of the above mentioned materials is similar for the operating range of the Re-Entrant fuelchannel. Tests have shown that the creep rate in Excel is much lower than Zr-2.5Nb, however Zirconium alloys are not suitable for long term exposure to supercritical water unless they are coated to avoid corrosion [9]. There is little data for irradiation creep and swelling at the high temperatures associated with the Re-Entrant fuel-channel for all of the above mentioned materials [6]. Extensive testing needs to be conducted using all four materials before the design of the Re-Entrant channel is finalized. For the purpose of this work, SS-304 has been chosen as the material of construction of the inner tube, the pressure tube and the outer tube. The inner tube in the Re-Entrant channel can be made as thin as possible to improve neutron economy as it is not required to bear any significant pressure difference[9]. A reference thickness of the inner tube is 2 mm, and a sensitivity analysis on the thickness of the inner tube is performed in Chapter 5. The inner diameter of the pressure tube is calculated to be mm. ASME standards require that the design stress of the pressure boundary component be less than 1/3 of the UTS of the material [9]. The UTS of SS-304 at 400 C is 448 MPa [36]. Analysis in Chapter 5 indicate that the actual temperature of the pressure tube is much lower than 400 C, but the UTS of SS-304 is chosen at this temperature to account for a reasonable safety factor. Thus, the minimum thickness of the pressure tube required to satisfy the ASME standards is mm. The reference pressure tube thickness for this work is chosen to be 11 mm. An insulator is required so that only 1 to 2 % of the thermal energy would be lost to the moderator from the Re-Entrant channel. Two different ceramic insulators are chosen in this work. The first is the ceramic insulator which is used in the HEC channel, so that heat loss comparisons can be made between the two designs. Thus, Porous Yttria 38

56 Stabilized Zirconia (with 70% porosity) has been chosen as for the insulation. Unlike the HEC channel, the Re-Entrant channel does not have SCW flowing through the pores of the insulator as the insulator is placed outside the pressure tube. Carbon dioxide (CO 2 ) gas flows through the pores in the insulator. Similar to current CANDU-type reactors, the moisture content of the CO 2 gas can be analyzed to determine if there is a leak in the pressure tube. Yttria Stabilized Zirconia (YSZ) has a low neutron cross section and low thermal conductivity. Studies have shown that irradiation would not significantly embrittle YSZ at high temperature [37]. Porous YSZ, with open pores increases the thermal resistance and improves the thermal shock resistance. The other insulating option selected for the Re-Entrant channel is solid zirconium dioxide (ZrO 2 ). The solid insulator eliminates the need for the CO 2 gas system. The reference thickness of the insulator is 7 mm and a sensitivity analysis on the insulator thickness can be seen in Chapter 5, along with a detailed comparison of the performance of both insulators selected for use in the Re-Entrant channel. The rear end of the Re-Entrant channel will be insulated in the end shield of the calandria. The main purpose of the outer tube is to prevent erosion of the ceramic insulator in the moderator. Hence, the outer tube can be as thin as possible to improve neutron economy and the reference thickness of the outer tube is 0.5 mm. Table 3.2 shows the relevant dimensions associated with the Re-Entrant channel and Figure 3.5 shows the crosssectional view of the channel. 39

57 Table 3.2: Reference case Re-Entrant Channel Dimensions Parameter Value Flow tube Inner Diameter (mm) Flow tube Outer Diameter (mm) Pressure tube ID (mm) Pressure tube OD (mm) Ceramic thickness (mm) 7 Outer tube ID (mm) Outer tube OD (mm) Notes: Flow tube, pressure tube, ceramic insulation and outer tube thicknesses may vary as a design option. Figure 3.5: Cross sectional view of Re-Entrant channel 40